Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc99m production
Peng Hong Liem, Hoai Nam Tran, Tagor Malem Sembiring, and Bakri Arbie
Citation: AIP Conference Proceedings 1615, 37 (2014); doi: 10.1063/1.4895857
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Conceptual Design of a New Homogeneous Reactor for
Medical Radioisotope Mo-99/Tc-99m Production
Peng Hong LIEM*, Hoai Nam TRAN§, Tagor Malem SEMBIRING¶ and Bakri
ARBIE†
*
Nippon Advanced Information Service (NAIS Co., Inc.) Scientific Computational Division,
416 Muramatsu, Tokaimura, Ibaraki, Japan
§
Chalmers University of Technology,Dept. of Applied Physics, Div. of Nuclear Engineering,
SE-412 96 Gothenburg, Sweden
¶
National Nuclear Energy Agency (BATAN), Center for Reactor Technology and Nuclear Safety, Kawasan
Puspiptek, Serpong, Tangerang Selatan, Banten, Indonesia
†
PT MOTAB Technology,
Kedoya Elok Plaza Blok DA 12, Jl. Panjang, Kebun Jeruk, Jakarta Barat, Indonesia
Abstract. To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuelsolution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The
modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for
decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters
are identified and the advantageous aspects are shown by comparing with the BATAN’s existing Mo-99 supply chain
which uses a heterogeneous reactor (RSG GAS multipurpose reactor).
INTRODUCTION
Since the end of 2007, shortages of Mo-99 supply occurred repeatedly due to several independent shutdowns
and outage extensions of major producing reactors [1, 2]. Since these major producing reactors, i.e. BR-2, HFR,
OSIRIS, NRU and SAFARI-I, covered 90 % to 95 % of the total global supply of Mo-99 (estimated to be 9,000 to
10,000 6-day Ci/week), the impact has been severe not only for developed countries but also for developing
countries. To secure the domestic (Indonesia) and regional (South East Asia) Mo-99 supply, there are at least two
options.
The first option is by increasing the production capacity of the existing BATAN’s Mo-99 supply chain. The
supply chain operated by BATAN consists of 30 MW Reaktor Serba Guna G.A. Siwabessy (RSG GAS)
multipurpose reactor for target irradiation, low enriched uranium (LEU) target preparation facility, processor facility
for Mo-99 extraction (based on the LEU modified Cintichem process [3] which has been acquired, licensed and
demonstrated commercially for decades), and Tc-99m generator manufacturing facility. Under the reported nominal
production capacity of 150 6-day Ci/week, BATAN is operating the facilities to meet the domestic demand as well
as export demand to Asian countries such as Vietnam, Bangladesh etc. The term “nominal” assumes the reactor is
operated at 30 MWth for 21 weeks in a year for multipurpose research and various irradiation activities.
In the present work, we conducted an assessment to increase the supply chain where all the irradiation positions
of RSG GAS are fully utilized while still meeting all safety and operating limits of the reactor and the coolability of
the LEU targets (however other irradiation activities may not be conducted). Under this maximum production mode,
the capacity can be maximized only up to 275 6-day Ci/week for 21 weeks in a year, i.e. just 1.8 times the nominal
capacity. The required U-235 for LEU targets and for fuel elements are 0.62 and 8.4 kg per year, respectively. The
LEU target power is around 83 kW and its ratio to reactor power is less than 0.3 % which shows the inefficient
utilization of U-235 and the resulting large amount of radioactive waste.
The second option is to introduce a new reactor dedicated for Mo-99 production and in the present work we
propose a nitrate-fuel-solution based homogeneous reactor for the following reasons: (1) no need of production and
transportation costs for LEU targets, (2) a reduction of fission product waste by several orders of magnitude for the
4th International Conference on Advances in Nuclear Science and Engineering (ICANSE 2013)
AIP Conf. Proc. 1615, 37-39 (2014); doi: 10.1063/1.4895857
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same specific activity of Mo-99/Tc-99m product, (3) a reduction of U-235 requirement of the same order compared
with the current research reactors, (4) minimal losses from Mo-99 decay by continuous extraction of Mo-99, (5)
variable reactor power based on demand and the possibility for modularity, and (6) high safety characteristics in
term of strong negative temperature and void reactivity of the core. In the next section, we report the on-going
conceptual design work, challenges and future works.
A NEW HOMOGENEOUS REACTOR DESIGN
Design Considerations and Main Parameters
The undergoing work on the conceptual design of a new homogeneous reactor has identified the following main
design considerations and parameters shown in Table 1.
TABLE 1. Main design parameters of the homogeneous reactor
Design Parameter
Value
Constraint
Thermal power (kW) per module
to be optimized
< 300
Power density (kW/L)
1–2
< 2.5
Fuel solution
UO 2 (NO 3 ) 2
–
U-235 enrichment (w/o)
19.75
< 20.0
Uranium concentration (gU/L)
200 – 300
< 500
Concentration of HNO 3 (mol/L)
0.1 – 0.5
–
Initial U-235 (kg)
3–5
–
Average fuel temperature (C)
80
< 100
As already mentioned before, in selecting the solution fuel for the homogeneous reactor we strongly considered
the matured technology for chemically processing the LEU targets (LEU modified Cintichem process).
Consequently, the UO 2 (NO 3 ) 2 (uranyl nitrate) based fuel solution is selected. This type of fuel solution has some
advantageous characteristics such as good solubility, easier isotope extraction and fuel solution purification.
Compared to the uranyl sulphate based fuel solution, the radiation stability of uranyl nitrate based solution is needed
attention since N 2 is produced during reactor operation in the rate of approximately 2.5 mL/kW.min [4].
The amount of U-235 required for initial loading and for subsequent compensation of U-235 depletion
determines the fuel cost of the reactor. As for the U-235 enrichment, the LEU, i.e. 19.75 w/o, is chosen for nuclear
security and proliferation reasons. The uranium concentration of the fuel solution was optimized and we found that
the minimal requirement for the initial critical loading would lie within 200 to 300 g U/L (cf. Figure 1). Within this
range of uranium concentration, the amount of U-235 for first criticality of the core is less than 3 kg. We also have
confirmed that, within this range, the temperature and void reactivity coefficients are all negative. All neutronics
analyses (criticality, kinetic parameters, feedback coefficients nuclear transmutation) were conducted using
SRAC2006 cell calculation code [5], MVP continuous energy Monte Carlo code [6] both with JENDL-3.3 based
libraries, DANTSYS neutron transport codes [7] and ORIGEN-2.2 code [8] with dedicated libraries to obtain
accurate nuclear transmutation results. The ORIGEN libraries were prepared using neutron spectra of the targeted
reactor where the preparation procedure is described in detail in Reference [9].
The estimated power density limit (2.5 kW/L) [4] is expected to be a major safety design constraint. In order to
achieve a certain production capacity, the reactor power may be freely determined to some extent by adjusting the
solution volume.
Comparison with RSG GAS (Heterogeneous Reactor)
The Mo-99 radioisotope production capacity of the homogeneous solution reactor depends almost linearly on the
reactor thermal power. As mentioned above, the thermal power and power density of the reactor are not fixed or
optimized yet in this stage but thermal hydraulic safety and operational constraints are expected to limit the two
parameters.
One example is given here for comparison with the previous mentioned RSG GAS under maximum production
mode where the reactor thermal power is set to about 83 kW in order to have the same production capacity. Our
analysis assumes that after Mo-99 isotope is extracted from the system, the reusable fuel solution is pumped back to
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the reactor after cleaning up process. It should be noted that the availability of the homogeneous reactor is also
assumed to be identical with RSG GAS (21 weeks in a year or approximately 40 %). For this availability factor, the
annual U-235 requirement for compensating U-235 depletion is slightly less than 10 g/y. For the same production
capacity of Mo-99 the RSG GAS as a heterogeneous reactor requires more than two order larger amount of U-235
for LEU targets not including the requirement for the fuel elements. In addition, the RSG GAS using LEU targets
produces the same order larger radioactive waste for the same production capacity, and almost all U-235 are
included in the waste. Not depending on the type of reactor, for this production capacity, in the Mo-99 processing
side, the present Mo-99 processor capacity in BATAN must be increased to 1.8 times of the present one.
4.5
7.0
4.0
6.0
3.5
Spectrum (1/lethargy)
Critical U-235 Mass (kg)
Uranium Concentration (250 gU/L)
8.0
5.0
Optimal
Range
4.0
3.0
2.0
1.0
200
300
2.5
2.0
1.5
Water Reflected (MVP)
1.0
Bare (SRAC)
0.5
0.0
100
3.0
400
Uranium Concentration (g U/L)
500
600
0.0
1.0E-05
1.0E-03
1.0E-01
1.0E+01
1.0E+03
1.0E+05
1.0E+07
Energy (eV)
FIGURE 1. Neutronics analyses results (left: critical U-235 mass for bare and water reflected cylindrical core where core
diameter is equal to core height; right: neutron spectrum at uranium concentration of 250 g U/L).
FUTURE WORKS
Challenging aspects of the undergoing conceptual design are summarized here: (a) on-power, in-core heat
removal system, (b) gas loop system for recombination of H 2 and O 2 to H 2 O, treatment of N 2 , fission product gases
etc, and (c) fuel solution purification system and nitric acid makeup system. Concerning the fuel solution
purification system, we pursue a system in which the fuel can be maximally reused after Mo-99 isotope is recovered
by separating fission products and if feasible other trans-uranium nuclides. Other foreseen issues are the economic
evaluation and the licensing of the design in accordance with the nuclear regulation in Indonesia.
REFERENCES
1. E. Parma, “The supply of the medical radioisotope Tc-99m/Mo-99 – Recent shortages call for action in developing a
domestic production capability,” American Nuclear Society, Trinity Section (November 2009).
2. OECD/NEA, “The supply of medical radioisotopes – The path to reliability,” Paris, French (2011).
3. Z. Aliludin et al., “Processing of LEU targets for Mo-99 production – Demonstration of a modified Cintichem process,” 1995
International Meeting on RERTR, Paris, French (September 1994).
4. IAEA, “Homogeneous aqueous solution nuclear reactors for the production of Mo-99 and other short lived radioisotopes,”
IAEA-TECDOC-1061, Vienna, Austria (2008).
5. K. Okumura et al., “SRAC2006: A comprehensive neutronics calculation code system,” JAEA-Data/Code 2007-004 (2007).
6. Y. Nagaya et al., “MVP/GMVP II: General purpose Monte Carlo codes for neutron and photon transport calculations based
on Continuous Energy and Multigroup Methods,” JAERI 1348 (2005).
7. R.E. Alcouffe et al., “DANTSYS: A diffusion accelerated neutron particle transport code system,” LA-12969-M, Los
Alamos National Laboratory (June 1995).
8. S.B. Ludwig and A.G. Croff, “Revision to ORIGEN2 – Version 2.2,” Transmittal Memo of CCC-0371/17, Oak Ridge
National Laboratory (2002).
9. P.H. Liem and T.M. Sembiring, “Development of new ORIGEN2 data library sets for research reactors with light water
cooled oxide and silicide LEU (20 w/o) fuels based on JENDL-3.3 nuclear data,” Nuclear Engineering and Design 262, pp.
52-62 (2013).
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